Corrosion Performance of Zircaloy-2 and Zircaloy-4 PWR Fuel Cladding

A variety of Zircaloy-2 and Zircaloy-4 cladding tubes with different microstructures has been manufactured. Test rods have been included in two fuel assemblies, and the assemblies have been inserted into the Ringhals 3 pressurized water reactor. Samples from the same cladding tube variants have been subjected to steam autoclave testing in the range 400 to 500°C for different exposure times and have been characterized with respect to their microstructure.

After one cycle of in-reactor exposure, the oxide thickness was thinner on stress relief annealed Zircaloy-2 rods than on Zircaloy-4 rods with the same heat treatment. Replacing the last stress relief anneal with a recrystallization anneal increased the out-of-pile corrosion resistance of the Zircaloy-4 cladding, whereas no statistical difference was seen in-pile after the first cycle. Scanning electron microscopy (SEM) measurements of the particle size distribution revealed only small differences in sizes between the cladding types.

The results from the autoclave tests done at 400°C/200 days, 410°C/28 days, 430°C/30 days, and 500°C/24 h indicated that the long-term 400°C test is capable of predicting the in-reactor behavior, whereas the 500°C as expected was found to be related to the nodular corrosion mechanism. The 430°C test seemed to be in a transition range from the uniform to the nodular corrosion mechanism. Short-term testing at 410°C resulted in the same ranking of materials as the long-term 400°C test indicating that a 410°C test could be used to predict the in-reactor behavior in a relatively short time.